Event Notification Report for May 26, 2000
U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
05/25/2000 - 05/26/2000
** EVENT NUMBERS **
36888 37031 37032 37033 37034
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 36888 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: WATERFORD REGION: 4 |NOTIFICATION DATE: 04/13/2000|
| UNIT: [3] [] [] STATE: LA |NOTIFICATION TIME: 17:13[EDT]|
| RXTYPE: [3] CE |EVENT DATE: 04/13/2000|
+------------------------------------------------+EVENT TIME: 15:35[CDT]|
| NRC NOTIFIED BY: E. LEMKE |LAST UPDATE DATE: 05/25/2000|
| HQ OPS OFFICER: JOHN MacKINNON +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |CHARLES PAULK R4 |
|10 CFR SECTION: | |
|AINC 50.72(b)(2)(iii)(C) POT UNCNTRL RAD REL | |
|AOUT 50.72(b)(1)(ii)(B) OUTSIDE DESIGN BASIS | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|3 N Y 100 Power Operation |100 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| FEEDWATER ISOLATION VALVES MAY CLOSE FASTER THAN 1.5-SECOND DESIGN BASIS |
| LIMIT. |
| |
| On March 21, 2000, Waterford Unit 3 determined that, based on a new |
| calculation methodology and the latest stroke time data, the Feedwater |
| Isolation Valves (FWIVs) FW-184 A (B) may close faster than the 1.5-second |
| design basis limit. |
| |
| The physical plant was determined to be operable and the FWIVs would have |
| performed their intended safety function at the time the condition was |
| identified. An initial operability evaluation was made in accordance with |
| procedure W4.101, which determined that the valves were operable at the time |
| the evaluation was conducted. This was based on an engineering evaluation |
| that determined that the faster closure of the FWIVs would not result in |
| water hammer loads that would prevent the FWIVs and their associated |
| penetrations from performing their required safety function. The |
| engineering evaluation determined that the analyzed increase in fast valve |
| closure (FVC) load is 57% for FW-184A and 52.9% for FW-184B. |
| |
| A subsequent evaluation was performed for FW-184A (B) to determine if at any |
| time in the last two years the increase in FVC load may have exceeded the |
| allowable loads determined by the engineering evaluation for W4.101. That |
| evaluation determined on one occasion for FW-184A and six occasions for |
| FW-184B, the percent increases for the FWIVs and subsequent stroke times |
| exceeded the values provided in the operability determination |
| |
| On these occasions in question, closure of FW-184A (B) in response to the |
| most adverse accident scenario could have potentially produced water hammer |
| that may have exceeded the capability of piping supports in the FW system |
| between the SGs and FW-184A (B). This may have resulted in the subsequent |
| loss of containment isolation function of FW-184A (B). |
| |
| The NRC Resident Inspector was notified of this event by the licensee. |
| |
| * * * UPDATE ON 05/25/00 AT 1630 ET BY CHRIS PICKERING TAKEN BY MACKINNON * |
| * * |
| |
| Entergy is revising this event notification. The event was initially |
| reported under 10CFR50.72(b)(2)(iii) as a condition that alone could have |
| prevented fulfillment of a safety function. Based upon further review of |
| the condition, Entergy has determined the condition is reportable under |
| 10CFR50.72(b)(1)(ii)(B) as a condition that was outside the design basis of |
| the plant. |
| |
| Evaluation of the event determined that the potential impact on containment |
| integrity as a result of this condition does not affect the results of the |
| currently analyzed MSLB event's radiological consequences. Further, the |
| possibility of containment integrity loss due to excessive waterhammer loads |
| during a FWLB remains bounded by the MSLB with respect to dose consequences. |
| The FWLB event radiological consequences are bounded by the MSLB event, |
| which is predicted to be within 10CFR100 limits. |
| |
| This event posed a containment integrity condition that was outside the |
| design basis of the plant. An evaluation determined that the FWIVs could |
| have potentially closed faster than their design basis minimum limit on |
| seven occasions, six instances for FW-184A and one instance for FW-184B. |
| The original event notification (above) incorrectly noted that there was one |
| instance for FW-184A and six instances for FW-184B. The faster valve |
| closure times could have placed increased loads on the piping and piping |
| supports in the FW system between the steam generators (inside containment) |
| and the FWIVs (outside containment). These increased loads were not |
| accounted for in the design basis of the system. Based on this information, |
| the event is reportable under 10CFR50.72(b)(1)(ii)(B) as a condition that |
| was outside the design basis of the plant. |
| |
| The NRC Resident Inspector was notified of the event revision by the |
| licensee. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 37031 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: CATAWBA REGION: 2 |NOTIFICATION DATE: 05/25/2000|
| UNIT: [1] [] [] STATE: SC |NOTIFICATION TIME: 09:21[EDT]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 05/25/2000|
+------------------------------------------------+EVENT TIME: 08:32[EDT]|
| NRC NOTIFIED BY: KEVIN PHILLIPS |LAST UPDATE DATE: 05/25/2000|
| HQ OPS OFFICER: FANGIE JONES +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |MARK LESSER R2 |
|10 CFR SECTION: | |
|AOUT 50.72(b)(1)(ii)(B) OUTSIDE DESIGN BASIS | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 100 Power Operation |100 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| AUXILIARY BUILDING FILTERED EXHAUST MAY NOT HAVE BEEN ABLE TO PERFORM ITS |
| DESIGN FUNCTION |
| |
| The licensee discovered both doors on the vestibule to the '1A' Centrifugal |
| Charging Pump Room not fully closed, this may have prevented the Auxiliary |
| Building filtered exhaust system from maintaining a negative pressure in |
| that room. This would render both trains of Auxiliary Building exhaust |
| inoperable per Technical Specifications 3.7.12 and would require entry into |
| Technical Specification 3.0.3. The doors were closed on discovery which |
| returned the unit to compliance with Technical Specifications. |
| |
| The licensee will followup with an investigation into when and how the doors |
| came to be open. |
| |
| The licensee notified the NRC Resident Inspector. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 37032 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: MCGUIRE REGION: 2 |NOTIFICATION DATE: 05/25/2000|
| UNIT: [1] [] [] STATE: NC |NOTIFICATION TIME: 21:47[EDT]|
| RXTYPE: [1] W-4-LP,[2] W-4-LP |EVENT DATE: 05/25/2000|
+------------------------------------------------+EVENT TIME: 20:46[EDT]|
| NRC NOTIFIED BY: H. M. HARRIS |LAST UPDATE DATE: 05/25/2000|
| HQ OPS OFFICER: JOHN MacKINNON +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |MARK LESSER R2 |
|10 CFR SECTION: |JOHN HANNON NRR |
|ARPS 50.72(b)(2)(ii) RPS ACTUATION |JOSEPH GIITTER IRO |
|AESF 50.72(b)(2)(ii) ESF ACTUATION | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 A/R Y 100 Power Operation |0 Hot Standby |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| TURBINE RUNBACK TO 50% DUE TO LOSS OF A MAIN FEEDWATER PUMP FOLLOWED BY A |
| REACTOR TRIP ON LOW LOW STEAM GENERATOR WATER LEVEL. ONE AUXILIARY |
| FEEDWATER PUMP HAD TO BE MANUALLY STARTED. |
| |
| Reactor automatic trip occurred at 2046 ET. Initiating event was steam |
| generator low low water level on "C" steam generator. Prior to the trip |
| Channel 1 vital 120 volt AC power was lost due to the tripping open of EVIA |
| (DC/AC ) inverter AC output breaker. No testing was in progress at this |
| time and the cause of the AC breaker trip is under investigation. When the |
| AC output breaker opened it caused a loss of Channel 1 power. Main |
| feedwater pump "1A" control circuitry interfaces with Channel 1 120 volt AC |
| power and when Main Feedwater pump "1A" control circuitry power was lost |
| main feedwater pump "1A" turbine tripped. The loss of Main Feedwater pump |
| "1A" turbine initiated an automatic main turbine runback to 50% power. |
| After the main turbine runback to 50% power the reactor tripped on steam |
| generator "1C" low low water level. Only one of two motor driven auxiliary |
| feedwater pumps automatically started on steam generator "1C" low low water |
| level. Auxiliary feedwater pump "1A" was manually started approximately 2 |
| minutes after the automatic reactor trip. The licensee is investigating why |
| the "1A" auxiliary feedwater pump did not automatically start. All rods |
| fully inserted into the core and reactor coolant temperature is being |
| maintained at Tave no load condition of 557 degrees F. No PORVs or code |
| safety valves on either the primary or secondary side of the plant opened. |
| All the Emergency Core Cooling Systems and the Emergency Diesel Generators |
| are fully operable if needed. Offsite electrical grid is stable. The |
| licensee's investigation into the initiating event is continuing. |
| |
| The NRC Resident Inspector was notified of this event by the licensee. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Fuel Cycle Facility |Event Number: 37033 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: PORTSMOUTH GASEOUS DIFFUSION PLANT |NOTIFICATION DATE: 05/25/2000|
| RXTYPE: URANIUM ENRICHMENT FACILITY |NOTIFICATION TIME: 23:52[EDT]|
| COMMENTS: 2 DEMOCRACY CENTER |EVENT DATE: 05/25/2000|
| 6903 ROCKLEDGE DRIVE |EVENT TIME: 18:00[EDT]|
| BETHESDA, MD 20817 (301)564-3200 |LAST UPDATE DATE: 05/26/2000|
| CITY: PIKETON REGION: 3 +-----------------------------+
| COUNTY: PIKE STATE: OH |PERSON ORGANIZATION |
|LICENSE#: GDP-2 AGREEMENT: N |JOHN MADERA R3 |
| DOCKET: 0707002 |WAYNE HODGES NMSS |
+------------------------------------------------+ |
| NRC NOTIFIED BY: JIM MCCLEERY | |
| HQ OPS OFFICER: FANGIE JONES | |
+------------------------------------------------+ |
|EMERGENCY CLASS: N/A | |
|10 CFR SECTION: | |
|NBNL RESPONSE-BULLETIN | |
| | |
| | |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| NRC BULLETIN 91-01, 24 HOUR NOTIFICATION |
| |
| The following is the faxed report from Portsmouth: |
| |
| On May 16, 2000, PORTS plant personnel initiated a review of an NRC event (# |
| 36993) submitted by Paducah Gaseous Diffusion Plant (PGDP), to assess |
| applicability at the Portsmouth sight. Errors associated with calibration |
| correction factors used to determine uranium masses were found to have not |
| occurred at the PORTS sight. |
| |
| On May 25. 2000, after further review by Nuclear Criticality Safety dept. |
| personnel, a violation of double contingency was identified when it was |
| determined that the independence of non-destructive assay (NDA) measurements |
| was not maintained. NCSA-PLANT013.A00 indicated that NDA measurements will |
| be maintained independent. However, calibration of instruments was not |
| maintained independent and therefore the resulting measurements were not |
| independent. As a result, the calibration process failed to ensure that a |
| single failure could not effect two independent mass measurements. Double |
| contingency was violated in cases where two independent uranium mass |
| measurements were required to establish double contingency. |
| |
| A review of NDA calibration data confirmed that no measurement errors |
| actually existed which would have affected the mass readings taken to |
| establish double contingency. This event is being reported because the |
| independence or mass measurement readings was not established and |
| maintained. |
| |
| SAFETY SIGNIFICANCE OF EVENTS: |
| |
| The safety significance of this event is very low. All historical |
| calibrations were reviewed (Ref. POEF-38.340.00.086) and no discrepancies |
| were identified with historical measurements. Furthermore, ongoing |
| laboratory Quality Assurance programs and practices help ensure errors of |
| this nature remain sufficiently unlikely to be relied on as criticality |
| controls. |
| |
| |
| POTENTIAL CRITICALITY PATHWAYS INVOLVED (BRIEF SCENARIO[S] OF HOW |
| CRITICALITY COULD OCCUR): |
| |
| An error in NDA measurements plus uncertainty would have to be low by |
| greater than a factor of two or results in a critical mass of uranium going |
| undetected. if this mass were then subject to the correct geometry, |
| moderation and reflection conditions a criticality could error |
| |
| |
| CONTROLLED PARAMETERS (MASS, MODERATION, GEOMETRY, CONCENTRATION, ETC.): |
| |
| Mass is the only controlled parameter. Two independent measurements of mass |
| are required to ensure double contingency. Since independence of NDA of NDA |
| mass measurements is called Into question, only one independent estimate of |
| mass is available. The absence of a second independent NDA mass measurement |
| represents a loss of one double contingency control. |
| |
| |
| ESTIMATED AMOUNT, ENRICHMENT, FORM OF LICENSED MATERIAL (INCLUDE PROCESS |
| LIMIT AND % WORST CASE OF CRITICAL MASS): |
| |
| No uranium was involved in the event. NDA measurements are used to classify |
| cascade deposits as either greater than safe mass or less than safe mass. |
| (Safe Mass is 43.5% of minimum critical mass.) NDA measurements are also |
| used in batching operations to ensure the 235U mass in the final container |
| is less than 350 grams 235U which is less half the minimum critical mass at |
| 100% enrichment and optimum moderation, geometry and reflection conditions. |
| |
| |
| NUCLEAR CRITICALITY SAFETY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION |
| OF THE FAILURES OR DEFICIENCIES |
| |
| Nuclear criticality safety controls include independent mass measurements of |
| 235U. Failure to ensure the calibration of the NDA instrumentation was |
| independent lead to a potential failure of the mass measurements. |
| |
| |
| CORRECTIVE ACTIONS TO RESTORE SAFETY SYSTEM AND WHEN EACH WAS IMPLEMENTED: |
| |
| NDA measurements performed for criticality safety purposes have been |
| suspended until independent verification of calibration data requirements is |
| flown into applicable procedures and programs. |
| |
| The certificate holder notified the NRC Resident Inspector and will notify |
| the DOE representative. |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
|Power Reactor |Event Number: 37034 |
+------------------------------------------------------------------------------+
+------------------------------------------------------------------------------+
| FACILITY: HOPE CREEK REGION: 1 |NOTIFICATION DATE: 05/26/2000|
| UNIT: [1] [] [] STATE: NJ |NOTIFICATION TIME: 02:29[EDT]|
| RXTYPE: [1] GE-4 |EVENT DATE: 05/25/2000|
+------------------------------------------------+EVENT TIME: 22:54[EDT]|
| NRC NOTIFIED BY: CHRIS SERATA |LAST UPDATE DATE: 05/26/2000|
| HQ OPS OFFICER: FANGIE JONES +-----------------------------+
+------------------------------------------------+PERSON ORGANIZATION |
|EMERGENCY CLASS: N/A |DAN HOLODY R1 |
|10 CFR SECTION: | |
|AIND 50.72(b)(2)(iii)(D) ACCIDENT MITIGATION | |
| | |
| | |
| | |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|UNIT |SCRAM CODE|RX CRIT|INIT PWR| INIT RX MODE |CURR PWR| CURR RX MODE |
+-----+----------+-------+--------+-----------------+--------+-----------------+
|1 N Y 15 Power Operation |15 Power Operation |
| | |
| | |
+------------------------------------------------------------------------------+
EVENT TEXT
+------------------------------------------------------------------------------+
| FOUR FILTRATION RECIRCULATION AND VENTILATION FANS TRIPPED UNEXPECTEDLY |
| |
| "On 5/25/00 at 22:54 hours, four operating Filtration, Recirculation. and |
| Ventilation System (FRVS) recirculation fans (A, B, D, E) unexpectedly |
| tripped while they were running for a monthly surveillance. An FRVS |
| ventilation fan, that was also in service, continued to run. |
| |
| "The FRVS consists of two subsystems, the Recirculation System and the |
| Ventilation System. The FRVS Recirculation System is an Engineered Safety |
| Feature (ESF) System, located inside the Secondary Containment, that reduces |
| offsite doses significantly below 10CFR100 guidelines during a LOCA, |
| refueling accident, or high radioactivity in the Secondary Containment. Upon |
| a Secondary Containment isolation, the FRVS Recirculation System is actuated |
| and recirculates the Secondary Containment air through filters for cleanup. |
| This subsystem is the initial cleanup system before discharge is made via |
| the FRVS Ventilation subsystem to the outdoors. The FRVS Ventilation System |
| is an ESF system, located inside the Secondary Containment, that maintains |
| the building at a negative pressure with respect to the outdoors. The system |
| takes suction from the discharge duct of the FRVS Recirculation system and |
| discharges the air through filters to the outdoors. |
| |
| "Investigation into the cause of the fan trips has identified a manual |
| damper in the ventilation system ductwork that failed to the closed |
| position. This manual damper is normally open during power operation. This |
| damper is repositioned closed during refueling outages to redistribute |
| ventilation through the Secondary Containment. Immediate actions were taken |
| to return the damper to the open position. Recirculation fans have been |
| returned to service and are operating satisfactorily. |
| |
| "The ventilation system ductwork was reviewed to identify other manual |
| dampers that could have the same or similar affect on system operation. Two |
| other dampers were identified, their positions have been verified to be |
| correct. |
| |
| "Hope Creek is presently in Operational Condition 1 at 15% power with the |
| turbine/generator off-line awaiting replacement of the 'C' Main Power |
| Transformer. All safety related equipment is available." |
| |
| The fans had been operating about 10 hours before tripping, it is suspected |
| that the damper was not adequately secured in the open position when the |
| plant exited the refueling outage. The licensee is investigating. |
| |
| The licensee notified the NRC Resident Inspector and the local township |
| authorities. |
+------------------------------------------------------------------------------+
Page Last Reviewed/Updated Thursday, March 25, 2021
Page Last Reviewed/Updated Thursday, March 25, 2021